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Effectiveness of source term optimization for higher disposal density of spent fuels in a deep geological repository
Annals of Nuclear Energy 71 (2014) 125–129
Contents lists available at ScienceDirect
Annals of Nuclear Energy journal homepage: www.elsevier.com/locate/anucene
Effectiveness of source term optimization for higher disposal density of spent fuels in a deep geological repository Dong-Keun Cho ⇑, Jongtae Jeong Korea Atomic Energy Research Institute, 150 Dukjin-dong, Yuseong, Daejeon 305-353, Republic of Korea
a r t i c l e
i n f o
Article history: Received 22 September 2013 Received in revised form 10 March 2014 Accepted 15 March 2014
Keywords: Spent fuel Disposal system Canister Disposal density Source term
a b s t r a c t Spent fuels to be disposed of have various initial 235U enrichments, discharge burnups, and cooling times, and consequently each type of spent fuel has a different decay heat. Because a disposal system design considering respective spent fuel characteristics was not previously possible in Korea, a system was developed based on a conservative reference spent fuel nominated through historical data analyses on the relation of spent fuel inventory with the initial 235U enrichment, discharge burnup, and cooling time. Recently, ASOURCE, an advanced source-term calculation system, was developed by the Korea Atomic Energy Research Institute, making it possible to design a disposal system considering the respective source terms of the spent fuels. In this study, the benefits resulting from a consideration of the individual decay heat for entire spent fuels were evaluated in terms of the footprint of a repository. As a result, it was seen that conservativeness in the disposal system design can be remarkably avoided, revealing a roughly 50% decrease in the footprint of the repository, considering the respective irradiation and cooling profile. It is therefore concluded that optimization of the source term is a viable option in a deep geological disposal system design. Ó 2014 Elsevier Ltd. All rights reserved.
1. Introduction For the efficient design of a disposal system for spent fuels (SFs), it is necessary to consider the exact decay heat, because it is one of the most important factors affecting the disposal area. The decay heat is a complicated function that is highly dependent on the discharge burnup and cooling time of the spent fuel. The discharge burnup depends on the initial 235U enrichment and neutronic features of the reactor in which the spent fuel was depleted. Even if the discharge burnup is the same, there are many spent fuels with different cooling times. It should be noted that if spent fuels have different discharge burnups or cooling times, they present a variety of source terms such as decay heat, radioactivity, nuclide concentration, and radiotoxicity. When the Korean Reference disposal System (KRS) was developed, a disposal system design considering the respective spent fuel characteristics was not possible in Korea due to the absence of a suitable code to estimate the source terms that takes into account each source term from entire spent fuels. Therefore, a disposal system was developed on the basis of conservative reference spent fuel determined by historical data analyses of the spent ⇑ Corresponding author. Tel.: +82 42 868 4899; fax: +82 42 868 8198. E-mail address: [email protected] (D.-K. Cho). http://dx.doi.org/10.1016/j.anucene.2014.03.018 0306-4549/Ó 2014 Elsevier Ltd. All rights reserved.
fuel inventory for the initial 235U enrichment, discharge burnup, and cooling time (Cho et al., 2007; Lee et al., 2007). Recently, an advanced source-term calculation system, called ASOURCE, was developed by the Korea Atomic Energy Research Institute. This code has a powerful capability to characterize the source terms considering respective fuel design, irradiation history, and cooling history. This means that a disposal system can be designed without inclusion of conservatism induced by unrealistic source term calculations for spent fuels (Cho et al., 2013a, 2013b). In this study, the benefits resulting from consideration of individual decay heat for entire spent fuels were evaluated in terms of the footprint of a deep geological repository to assess the viability of source term optimization in the disposal system design.
2. Description of Korean reference disposal system The KRS was designed to accommodate the spent fuels from Korea’s 4 CANDU and 16 PWR reactors. In this paper, only a disposal system for PWR spent fuel is introduced, because the effectiveness of the source term optimization was investigated only for a disposal system for PWR spent fuel.
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2.1. Source terms and inventory
Unit: cm
As described earlier, when the KRS was developed, source terms were estimated for the reference spent fuel determined based on a statistical analysis of the initial 235U enrichment, discharge burnup, and cooling time discharged by the end of 2006. The reference fuel has an initial 235U enrichment of 4.0 wt.%, discharge burnup of 45,000 MWd/tU, and cooling time of 40 years (Cho et al., 2007; Lee et al., 2007). The regression equation of decay heat for the reference spent fuel is given in Eq. (1), which was used as input data for a thermal-mechanical analysis for a disposal system design.
PðtÞ ¼ 4545:68 t 0:75756 ½W=tHM ð30 6 t 6 106 yearsÞ
ð1Þ
According to Eq. (1), the decay heat from the reference spent fuel is approximately 890 W/tHM at 40 years after a reactor discharge, which is equivalent to a 1.6 kW/canister. The total amount of spent fuel, 20,000 MtU, expected from the 16 PWR reactors was considered as a design basis when the KRS was developed. 2.2. The disposal system 2.2.1. General site assumption Since a repository site for spent fuel has not been discussed in Korea, the design of the KRS was performed on the basis of geological data on granite, which is considered a candidate rock for a disposal system in the R&D discipline (Korea Atomic Energy Research Institute, 2011). It was assumed that the repository would be constructed in a granite rock foundation. A geothermal gradient of 30 °C/km was used in the thermal analysis (Lee et al., 2007). 2.2.2. Disposal canister and buffer The main role of the canister is to confine radionuclides during the specified period. The canisters consist of two parts, an outer shell made of copper for corrosion resistance and an insert made of nodular cast iron for structural integrity. The thickness of the outer shell is 5 cm. Fig. 1 shows a disposal canister called KDC-1 to accommodate PWR spent fuel. A conceptual design of the KRS was proposed based on the KDC-1, which contains four PWR spent fuel assemblies. The insert was designed to withstand the hydrostatic pressure and swelling pressure caused by the bentonite buffer. The dimensions for the KDC-1 canister are summarized in Fig. 1. In a geological disposal system, one of the major roles of a buffer is to protect the disposal canister under the given geological conditions. Generally, the buffer material is introduced to prevent seepage of groundwater into the disposal canister. Domestic Korean Ca-bentonite with a dry density of 1.6 g/cm3, which gives a thermal conductivity of 1.0 W/m °C, was chosen as a candidate buffer material for the KRS. The thickness of the buffer was determined to be 50 cm to provide adequate retardation and adsorption of the radionuclides. Fig. 2 shows the configuration of the disposal hole, including the buffer and canister. 2.2.3. Layout of the disposal system The depth of the repository is set to be 500 m below the surface. The repository consists of three sections: a disposal area, technical rooms in the controlled area, and technical rooms in the uncontrolled area. The disposal area consists of disposal tunnels, panel tunnels, and a central tunnel. Fig. 3 shows the configuration of the deposition tunnel and holes. The distance between the deposition holes was determined through a thermal analysis. The constraint of the peak temperature in the buffer was 100 °C. The buffer should be kept below the limit temperature to maintain per-
Fig. 1. Schematic illustration of a KDC-1 canister.
formance of swelling pressure. The spacing of 40 m for tunnels and a distance of 6 m between deposition holes satisfied the contraints, revealing a peak temperature of 98.6 °C. For more detailed information, please refer to a study by Lee et al. (2007).
3. Effectiveness of source term optimization 3.1. Source term evaluation tool ORIGEN2 (Croff, 1980) or ORIGEN-ARP (Gauld et al., 2009) is generally used to estimate the source term needed for the facility design for the storage or disposal of spent fuels. These codes can only estimate the single irradiation and decay history for a specified assembly design. Therefore, a disposal system design considering the respective source terms of entire spent fuels is impossible with these codes. However, an advanced source-term evaluation program, called ASOURCE, was developed by the Korea Atomic Energy Research Institute for a source term analysis to accomplish the advanced fuel cycle being considered in the Republic of Korea. ASOURCE has the following features: (a) estimation of inflow and outflow source terms in each unit process for the design of the pyroprocess facility; (b) overall inventory calculation for long-lived nuclides and transuranics in SFs stored at each or all reactor sites; and (c) evaluation of grand source terms for a batch of SFs with different irradiation and cooling characteristics to avoid conservative design of a storage or disposal facility. Currently, functional modules, called Screening, DeplDec, DecRes, ReproRun, MetalRun, and Batch, are also available within the system. The Screening module selects SFs, for which the calculations are to be done, from a SF database according to a user definition. The SF database includes the fuel design characteristics, irradiation
D.-K. Cho, J. Jeong / Annals of Nuclear Energy 71 (2014) 125–129
Fig. 2. Schematic illustration of a buffer and canister for spent fuels.
127
characteristics, and cooling characteristics for each fuel assembly identification (ID). The role of the DeplDec module is to perform depletion and decay calculations with appropriate physics parameters. The role of the DecRes module is to accomplish a decay calculation for SF or radwaste by a restart calculation with the precalculated data. The role of the ReproRun module is to characterize radwastes generated from the pyroprocess considering the material balance of each nuclide for each unit process. The role of MetalRun is to carry out depletion and decay calculations for the structural materials of the assembly hardware. The role of the Batch module is to estimate the mixture composition when a variety of SFs or structural components are processed together at the same time, t. ASOURCE currently has three analysis sequences, a Fuel Waste Characterization Sequence, Metal Waste Characterization Sequence, and Grand Source Term Characterization Sequence. By calling the appropriate functional modules outlined above, it can carry out user-defined tasks. Refer to studies by Cho et al. (2013a, 2013b) for a more detailed explanation. In this study, the Grand Source Term Characterization Sequence of the ASOURCE was used to characterize average source terms of spent fuels resulting from respective depletion and decay calculations. The remarkable feature of this sequence is its capability to calculate the average source terms for a group of SFs corresponding to the user-specified bin. It is operated as follows. In the first step, the Screening module chooses SFs according to the user-supplied bin of 235U enrichment, discharge burunup, and discharge year from t0 to t1, resulting in group 1, followed by the individual depletion and decay calculations by the DeplDec module for the selected assemblies up to the end of the calendar year, t1. The Batch module then makes the mixture composition, resulting in mixture 1, taking into account the initial uranium loading, the mixing ratio, and the nuclide composition vector at the final end of the calendar year of t1. Finally, mixture 1 is decayed by the DecRes module up to the final date of the calendar year of t2. In the second step, the Screening module searches SFs appropriate to the user-defined bin for discharge year from t1 to t2, resulting in group 2, followed by the
Fig. 3. Concept of a disposal canister and an engineering barrier of KRS.
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respective depletion and decay calculations for the selected assemblies up to the end of the calendar year of t2. The Batch module then calculates the mixture composition, mixture 2, by combining the nuclide composition of mixture 1 at the end of the calendar year of t2 and the respective nuclide composition of assembly in group 2 at the end of the calendar year of t2. Finally, a decay calculation of mixture 2 is performed by the DecRes module up to the end of the calendar year of t3. By repeating the above procedure, the average source terms as a function of calendar year from t1 to tn are estimated. If information on SFs to be discharged in the future according to a national plan is added to the SF database, the average source terms can be evaluated as a function of the upcoming time. Refer to (Cho et al., 2013a) for a more detailed explanation. 3.2. Source term optimization Fig. 4. Decay heat for reference fuel and spent fuels at Kori unit 4.
The main purpose of this paper is to investigate whether optimization of the source terms could be an important parameter in the disposal system design. The goal of the optimization is to decrease the disposal hole spacing with satisfying the maximum temperature of 100 °C of buffer for minimization of footprint of a repository. Table 1 shows the regression formula for the predicted value of decay heat for the reference SF with the exact information-based value for SFs stored at Kori unit 4. Because the reference fuel was determined by analyzing spent fuel information stockpiled by 2005 (Cho et al., 2006), it is almost equivalent to comparing the decay heat from the entire SFs discharged from all reactor sites with that discharged from all SFs stockpiled at Kori unit 4 by the end of 2007. These SFs revealed an initial enrichment distribution of 1.6–4.5 wt.%, and a discharge burnup of 12,000–54,000 MWd/ MTU, yielding 960 assemblies. Fig. 4 shows the comparison results for the decay heat of the reference fuel used for the design of the KRS and the average source terms of the entire SFs discharged from Kori unit 4 until 2007. Because the reference fuel for the KRS was determined by analyzing the spent fuel information stockpiled by 2005, it is meaningful to compare the decay heat from the reference fuel with that from all SFs stockpiled at Kori unit 4. In the figure, the dotted line denotes the decay heat of the reference fuel, and the solid line represents the average decay heat calculated by ASOURCE considering the respective irradiation and decay profiles for 960 assemblies. As shown in the figure, approximately 1.5 times higher decay heat was applied for the KRS accommodating SFs discharged from Kori unit 4. 3.3. Effectiveness of source term optimization A thermal analysis for KRS using ABAQUS 6-10 was performed with the average decay heat calculated by ASOURE, as listed in Table 1. The temperature profile for the KRS as a function of time was analyzed with the transient mode.
Fig. 5. Thermal analysis model.
As shown in Fig. 5, the depth from the surface to 1000 m was considered in the model, including the disposal tunnel and bentonite buffer surrounding the canister. One-fourth of the model was simulated by applying adiabatic boundary conditions at two midplanes. The initial temperature was set to 10 °C at the ground surface and 40 °C at 1000 m, which is the bottom surface of the model. A geothermal gradient of 30 °C/km was also used in the thermal analysis. Fig. 6(a) represents the results of a sensitivity study on the peak temperature of a buffer. The change of temperature is more sensitive to a borehole spacing than tunnel spacing, resulting in remarkable change of repository area. From this analysis, it was shown that the new disposal system design with a tunnel spacing of
Table 1 Regression formula and coefficients for decay heat. Time (years)
Equation
1–100
Y ¼ y0 þ A1 e
100–106
Constant ðtx Þ 1
ðtx Þ
Y ¼ y0 þ A1 e
1
ðtx Þ
þ A2 e
2
ðtx Þ
þ A2 e
2
þ A3 e
ðtx Þ 3
y0
138.4
A1 A2 A3
286.5 1041.5 780.5
y0
22.3
A1 A2
162.3 412.2
t1 t2 t3
22.004 1.646 61.144
t1 t2
684.61 87.216
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7
100
45
90 80
5
35
4
30
70
o
40
Temperature ( C)
6
Tunnel spacing (m)
Bore hole spacing (m)
Fixed tunnel spacing : 30m Fixed hole spacing : 4m
60 50 40 30 20
New_PWR_30m-4m
10 3 70
75
80
85
90
95
25 100
Old_PWR_40m-6m
0 0
200
o
400
600
800
1000
Time (years)
Temperature ( C)
(a) with source terms from ASOURCE
(b) for two proposed designs
Fig. 6. Temperature as a function of time for various conceptual disposal systems.
30 m and a borehole spacing of 4 m can satisfy the temperature constraint, resulting in a peak temperature of 94.5 °C. Fig. 6(b) shows comparison results on the peak temperature as a function of time for the two proposed designs. The peak temperatures of optimized design proposed in this study and the KRS design revealed at 41 and 15 years after deposition, respectively. Because decay heat reduction of the reference spent fuel is steeper than that of spent fuel proposed in this study, earlier peak satisfying temperature constraint and rapid decrease of temperature at buffer were induced in the KRS design. As shown in the figure, the newly proposed design with an average decay heat can decrease the footprint of a repository by up to about 50%. By applying the average source term in the disposal system design, spent fuel with a high burnup or short cooling time can be mixed with spent fuel with a low burnup or long cooling time. This indicates that optimization of the source term can decrease the area of a repository remarkably, and the application of average source term is an efficient way in a deep geological disposal system design. In this paper, effectiveness of source term optimization was judged on the basis of spent fuels discharged from Kori unit 4. If entire spent fuels from all life-time of existing reactors are considered, the quantitative benefit will be changed. This conclusion, however, will not be changed, because decay heat considering burnup and cooling time of respective spent fuel will be definitely lower than that of reference spent fuel with conservativeness. 4. Conclusions Spent fuels to be disposed of have various initial 235U enrichments, discharge burnups, and cooling times, and consequently each spent fuel has a different decay heat. Because a disposal system design considering the respective spent fuel characteristics was previously not possible in Korea, a system was developed based on a conservative reference spent fuel nominated through historical data analyses on the relation of the spent fuel inventory with the initial 235U enrichment, discharge burnup, and cooling
time. Recently, ASOURCE, an advanced source-term calculation system, was developed by the Korea Atomic Energy Research Institute, making it possible to design a disposal system considering the respective source terms of spent fuels. In this study, the benefits resulting from consideration of the individual decay heat for entire spent fuels were evaluated in terms of the footprint of a repository. As a result, it was seen that conservativeness in the disposal system design can be remarkably avoided, revealing about a roughly 50% decrease in the footprint of repository, considering the respective irradiation and cooling profile. Therefore, it was concluded that optimization of the source term is a viable option in a deep geological disposal system design. Acknowledgements We would like to acknowledge that this work was funded by the Ministry of Science, ICT and Future Planning. References Cho, D.K., Choi, J.W., Hahn, P.S., 2006. Current status and projection of spent nuclear fuel for geological disposal system design. J. Kor. Radioactive Waste Soc. 4 (1), 87–93. Cho, D.K., Lee, Y., Lee, J.Y., Choi, J.W., 2007. Characteristics of a geological disposal system for the increasing burn-up of spent nuclear fuel in Korea. J. Nucl. Sci. Technol. 44 (10), 1306–1316. Cho, D.K., Kook, D.H., Choi, J.W., Park, J.H., 2013a. Advanced hybrid analysis system for nuclear facility design with best estimate source terms. Nucl. Eng. Des. 256 (2013), 274–284. Cho, D.K., Choi, H.J., Jeong, J., 2013b. Verification of source term estimation method against measured data for spent fuel hardware characterization. Ann. Nucl. Energy 58 (2013), 36–42. Croff, G., 1980. A User’s Manual for the ORIGEN2 Computer Code, ORNL/TM-7175, Oak Ridge National Laboratory Report. Gauld, I.C., Bowman, S.M., Horwedel, J.E., 2009. ORIGEN-ARP: Automatic Rapid Processing for Spent Fuel Depletion, Decay, And Source Term Analysis, ORNL/ TM-2005/39, Version 6, Vol. I, Section, D1. Oak Ridge National Laboratory. Korea Atomic Energy Research Institute, 2011. Geological Disposal of Pyroprocessed Waste from PWR Spent Nuclear Fuel in Korea. KAERI/TR-4525/2011. Lee, J.Y., Cho, D.K., Choi, H.J., Choi, J.W., 2007. Concept of a Korean reference disposal system for spent fuels. J. Nucl. Sci. Technol. 44 (12), 1565–1573.
Contents lists available at ScienceDirect
Annals of Nuclear Energy journal homepage: www.elsevier.com/locate/anucene
Effectiveness of source term optimization for higher disposal density of spent fuels in a deep geological repository Dong-Keun Cho ⇑, Jongtae Jeong Korea Atomic Energy Research Institute, 150 Dukjin-dong, Yuseong, Daejeon 305-353, Republic of Korea
a r t i c l e
i n f o
Article history: Received 22 September 2013 Received in revised form 10 March 2014 Accepted 15 March 2014
Keywords: Spent fuel Disposal system Canister Disposal density Source term
a b s t r a c t Spent fuels to be disposed of have various initial 235U enrichments, discharge burnups, and cooling times, and consequently each type of spent fuel has a different decay heat. Because a disposal system design considering respective spent fuel characteristics was not previously possible in Korea, a system was developed based on a conservative reference spent fuel nominated through historical data analyses on the relation of spent fuel inventory with the initial 235U enrichment, discharge burnup, and cooling time. Recently, ASOURCE, an advanced source-term calculation system, was developed by the Korea Atomic Energy Research Institute, making it possible to design a disposal system considering the respective source terms of the spent fuels. In this study, the benefits resulting from a consideration of the individual decay heat for entire spent fuels were evaluated in terms of the footprint of a repository. As a result, it was seen that conservativeness in the disposal system design can be remarkably avoided, revealing a roughly 50% decrease in the footprint of the repository, considering the respective irradiation and cooling profile. It is therefore concluded that optimization of the source term is a viable option in a deep geological disposal system design. Ó 2014 Elsevier Ltd. All rights reserved.
1. Introduction For the efficient design of a disposal system for spent fuels (SFs), it is necessary to consider the exact decay heat, because it is one of the most important factors affecting the disposal area. The decay heat is a complicated function that is highly dependent on the discharge burnup and cooling time of the spent fuel. The discharge burnup depends on the initial 235U enrichment and neutronic features of the reactor in which the spent fuel was depleted. Even if the discharge burnup is the same, there are many spent fuels with different cooling times. It should be noted that if spent fuels have different discharge burnups or cooling times, they present a variety of source terms such as decay heat, radioactivity, nuclide concentration, and radiotoxicity. When the Korean Reference disposal System (KRS) was developed, a disposal system design considering the respective spent fuel characteristics was not possible in Korea due to the absence of a suitable code to estimate the source terms that takes into account each source term from entire spent fuels. Therefore, a disposal system was developed on the basis of conservative reference spent fuel determined by historical data analyses of the spent ⇑ Corresponding author. Tel.: +82 42 868 4899; fax: +82 42 868 8198. E-mail address: [email protected] (D.-K. Cho). http://dx.doi.org/10.1016/j.anucene.2014.03.018 0306-4549/Ó 2014 Elsevier Ltd. All rights reserved.
fuel inventory for the initial 235U enrichment, discharge burnup, and cooling time (Cho et al., 2007; Lee et al., 2007). Recently, an advanced source-term calculation system, called ASOURCE, was developed by the Korea Atomic Energy Research Institute. This code has a powerful capability to characterize the source terms considering respective fuel design, irradiation history, and cooling history. This means that a disposal system can be designed without inclusion of conservatism induced by unrealistic source term calculations for spent fuels (Cho et al., 2013a, 2013b). In this study, the benefits resulting from consideration of individual decay heat for entire spent fuels were evaluated in terms of the footprint of a deep geological repository to assess the viability of source term optimization in the disposal system design.
2. Description of Korean reference disposal system The KRS was designed to accommodate the spent fuels from Korea’s 4 CANDU and 16 PWR reactors. In this paper, only a disposal system for PWR spent fuel is introduced, because the effectiveness of the source term optimization was investigated only for a disposal system for PWR spent fuel.
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2.1. Source terms and inventory
Unit: cm
As described earlier, when the KRS was developed, source terms were estimated for the reference spent fuel determined based on a statistical analysis of the initial 235U enrichment, discharge burnup, and cooling time discharged by the end of 2006. The reference fuel has an initial 235U enrichment of 4.0 wt.%, discharge burnup of 45,000 MWd/tU, and cooling time of 40 years (Cho et al., 2007; Lee et al., 2007). The regression equation of decay heat for the reference spent fuel is given in Eq. (1), which was used as input data for a thermal-mechanical analysis for a disposal system design.
PðtÞ ¼ 4545:68 t 0:75756 ½W=tHM ð30 6 t 6 106 yearsÞ
ð1Þ
According to Eq. (1), the decay heat from the reference spent fuel is approximately 890 W/tHM at 40 years after a reactor discharge, which is equivalent to a 1.6 kW/canister. The total amount of spent fuel, 20,000 MtU, expected from the 16 PWR reactors was considered as a design basis when the KRS was developed. 2.2. The disposal system 2.2.1. General site assumption Since a repository site for spent fuel has not been discussed in Korea, the design of the KRS was performed on the basis of geological data on granite, which is considered a candidate rock for a disposal system in the R&D discipline (Korea Atomic Energy Research Institute, 2011). It was assumed that the repository would be constructed in a granite rock foundation. A geothermal gradient of 30 °C/km was used in the thermal analysis (Lee et al., 2007). 2.2.2. Disposal canister and buffer The main role of the canister is to confine radionuclides during the specified period. The canisters consist of two parts, an outer shell made of copper for corrosion resistance and an insert made of nodular cast iron for structural integrity. The thickness of the outer shell is 5 cm. Fig. 1 shows a disposal canister called KDC-1 to accommodate PWR spent fuel. A conceptual design of the KRS was proposed based on the KDC-1, which contains four PWR spent fuel assemblies. The insert was designed to withstand the hydrostatic pressure and swelling pressure caused by the bentonite buffer. The dimensions for the KDC-1 canister are summarized in Fig. 1. In a geological disposal system, one of the major roles of a buffer is to protect the disposal canister under the given geological conditions. Generally, the buffer material is introduced to prevent seepage of groundwater into the disposal canister. Domestic Korean Ca-bentonite with a dry density of 1.6 g/cm3, which gives a thermal conductivity of 1.0 W/m °C, was chosen as a candidate buffer material for the KRS. The thickness of the buffer was determined to be 50 cm to provide adequate retardation and adsorption of the radionuclides. Fig. 2 shows the configuration of the disposal hole, including the buffer and canister. 2.2.3. Layout of the disposal system The depth of the repository is set to be 500 m below the surface. The repository consists of three sections: a disposal area, technical rooms in the controlled area, and technical rooms in the uncontrolled area. The disposal area consists of disposal tunnels, panel tunnels, and a central tunnel. Fig. 3 shows the configuration of the deposition tunnel and holes. The distance between the deposition holes was determined through a thermal analysis. The constraint of the peak temperature in the buffer was 100 °C. The buffer should be kept below the limit temperature to maintain per-
Fig. 1. Schematic illustration of a KDC-1 canister.
formance of swelling pressure. The spacing of 40 m for tunnels and a distance of 6 m between deposition holes satisfied the contraints, revealing a peak temperature of 98.6 °C. For more detailed information, please refer to a study by Lee et al. (2007).
3. Effectiveness of source term optimization 3.1. Source term evaluation tool ORIGEN2 (Croff, 1980) or ORIGEN-ARP (Gauld et al., 2009) is generally used to estimate the source term needed for the facility design for the storage or disposal of spent fuels. These codes can only estimate the single irradiation and decay history for a specified assembly design. Therefore, a disposal system design considering the respective source terms of entire spent fuels is impossible with these codes. However, an advanced source-term evaluation program, called ASOURCE, was developed by the Korea Atomic Energy Research Institute for a source term analysis to accomplish the advanced fuel cycle being considered in the Republic of Korea. ASOURCE has the following features: (a) estimation of inflow and outflow source terms in each unit process for the design of the pyroprocess facility; (b) overall inventory calculation for long-lived nuclides and transuranics in SFs stored at each or all reactor sites; and (c) evaluation of grand source terms for a batch of SFs with different irradiation and cooling characteristics to avoid conservative design of a storage or disposal facility. Currently, functional modules, called Screening, DeplDec, DecRes, ReproRun, MetalRun, and Batch, are also available within the system. The Screening module selects SFs, for which the calculations are to be done, from a SF database according to a user definition. The SF database includes the fuel design characteristics, irradiation
D.-K. Cho, J. Jeong / Annals of Nuclear Energy 71 (2014) 125–129
Fig. 2. Schematic illustration of a buffer and canister for spent fuels.
127
characteristics, and cooling characteristics for each fuel assembly identification (ID). The role of the DeplDec module is to perform depletion and decay calculations with appropriate physics parameters. The role of the DecRes module is to accomplish a decay calculation for SF or radwaste by a restart calculation with the precalculated data. The role of the ReproRun module is to characterize radwastes generated from the pyroprocess considering the material balance of each nuclide for each unit process. The role of MetalRun is to carry out depletion and decay calculations for the structural materials of the assembly hardware. The role of the Batch module is to estimate the mixture composition when a variety of SFs or structural components are processed together at the same time, t. ASOURCE currently has three analysis sequences, a Fuel Waste Characterization Sequence, Metal Waste Characterization Sequence, and Grand Source Term Characterization Sequence. By calling the appropriate functional modules outlined above, it can carry out user-defined tasks. Refer to studies by Cho et al. (2013a, 2013b) for a more detailed explanation. In this study, the Grand Source Term Characterization Sequence of the ASOURCE was used to characterize average source terms of spent fuels resulting from respective depletion and decay calculations. The remarkable feature of this sequence is its capability to calculate the average source terms for a group of SFs corresponding to the user-specified bin. It is operated as follows. In the first step, the Screening module chooses SFs according to the user-supplied bin of 235U enrichment, discharge burunup, and discharge year from t0 to t1, resulting in group 1, followed by the individual depletion and decay calculations by the DeplDec module for the selected assemblies up to the end of the calendar year, t1. The Batch module then makes the mixture composition, resulting in mixture 1, taking into account the initial uranium loading, the mixing ratio, and the nuclide composition vector at the final end of the calendar year of t1. Finally, mixture 1 is decayed by the DecRes module up to the final date of the calendar year of t2. In the second step, the Screening module searches SFs appropriate to the user-defined bin for discharge year from t1 to t2, resulting in group 2, followed by the
Fig. 3. Concept of a disposal canister and an engineering barrier of KRS.
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D.-K. Cho, J. Jeong / Annals of Nuclear Energy 71 (2014) 125–129
respective depletion and decay calculations for the selected assemblies up to the end of the calendar year of t2. The Batch module then calculates the mixture composition, mixture 2, by combining the nuclide composition of mixture 1 at the end of the calendar year of t2 and the respective nuclide composition of assembly in group 2 at the end of the calendar year of t2. Finally, a decay calculation of mixture 2 is performed by the DecRes module up to the end of the calendar year of t3. By repeating the above procedure, the average source terms as a function of calendar year from t1 to tn are estimated. If information on SFs to be discharged in the future according to a national plan is added to the SF database, the average source terms can be evaluated as a function of the upcoming time. Refer to (Cho et al., 2013a) for a more detailed explanation. 3.2. Source term optimization Fig. 4. Decay heat for reference fuel and spent fuels at Kori unit 4.
The main purpose of this paper is to investigate whether optimization of the source terms could be an important parameter in the disposal system design. The goal of the optimization is to decrease the disposal hole spacing with satisfying the maximum temperature of 100 °C of buffer for minimization of footprint of a repository. Table 1 shows the regression formula for the predicted value of decay heat for the reference SF with the exact information-based value for SFs stored at Kori unit 4. Because the reference fuel was determined by analyzing spent fuel information stockpiled by 2005 (Cho et al., 2006), it is almost equivalent to comparing the decay heat from the entire SFs discharged from all reactor sites with that discharged from all SFs stockpiled at Kori unit 4 by the end of 2007. These SFs revealed an initial enrichment distribution of 1.6–4.5 wt.%, and a discharge burnup of 12,000–54,000 MWd/ MTU, yielding 960 assemblies. Fig. 4 shows the comparison results for the decay heat of the reference fuel used for the design of the KRS and the average source terms of the entire SFs discharged from Kori unit 4 until 2007. Because the reference fuel for the KRS was determined by analyzing the spent fuel information stockpiled by 2005, it is meaningful to compare the decay heat from the reference fuel with that from all SFs stockpiled at Kori unit 4. In the figure, the dotted line denotes the decay heat of the reference fuel, and the solid line represents the average decay heat calculated by ASOURCE considering the respective irradiation and decay profiles for 960 assemblies. As shown in the figure, approximately 1.5 times higher decay heat was applied for the KRS accommodating SFs discharged from Kori unit 4. 3.3. Effectiveness of source term optimization A thermal analysis for KRS using ABAQUS 6-10 was performed with the average decay heat calculated by ASOURE, as listed in Table 1. The temperature profile for the KRS as a function of time was analyzed with the transient mode.
Fig. 5. Thermal analysis model.
As shown in Fig. 5, the depth from the surface to 1000 m was considered in the model, including the disposal tunnel and bentonite buffer surrounding the canister. One-fourth of the model was simulated by applying adiabatic boundary conditions at two midplanes. The initial temperature was set to 10 °C at the ground surface and 40 °C at 1000 m, which is the bottom surface of the model. A geothermal gradient of 30 °C/km was also used in the thermal analysis. Fig. 6(a) represents the results of a sensitivity study on the peak temperature of a buffer. The change of temperature is more sensitive to a borehole spacing than tunnel spacing, resulting in remarkable change of repository area. From this analysis, it was shown that the new disposal system design with a tunnel spacing of
Table 1 Regression formula and coefficients for decay heat. Time (years)
Equation
1–100
Y ¼ y0 þ A1 e
100–106
Constant ðtx Þ 1
ðtx Þ
Y ¼ y0 þ A1 e
1
ðtx Þ
þ A2 e
2
ðtx Þ
þ A2 e
2
þ A3 e
ðtx Þ 3
y0
138.4
A1 A2 A3
286.5 1041.5 780.5
y0
22.3
A1 A2
162.3 412.2
t1 t2 t3
22.004 1.646 61.144
t1 t2
684.61 87.216
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D.-K. Cho, J. Jeong / Annals of Nuclear Energy 71 (2014) 125–129
7
100
45
90 80
5
35
4
30
70
o
40
Temperature ( C)
6
Tunnel spacing (m)
Bore hole spacing (m)
Fixed tunnel spacing : 30m Fixed hole spacing : 4m
60 50 40 30 20
New_PWR_30m-4m
10 3 70
75
80
85
90
95
25 100
Old_PWR_40m-6m
0 0
200
o
400
600
800
1000
Time (years)
Temperature ( C)
(a) with source terms from ASOURCE
(b) for two proposed designs
Fig. 6. Temperature as a function of time for various conceptual disposal systems.
30 m and a borehole spacing of 4 m can satisfy the temperature constraint, resulting in a peak temperature of 94.5 °C. Fig. 6(b) shows comparison results on the peak temperature as a function of time for the two proposed designs. The peak temperatures of optimized design proposed in this study and the KRS design revealed at 41 and 15 years after deposition, respectively. Because decay heat reduction of the reference spent fuel is steeper than that of spent fuel proposed in this study, earlier peak satisfying temperature constraint and rapid decrease of temperature at buffer were induced in the KRS design. As shown in the figure, the newly proposed design with an average decay heat can decrease the footprint of a repository by up to about 50%. By applying the average source term in the disposal system design, spent fuel with a high burnup or short cooling time can be mixed with spent fuel with a low burnup or long cooling time. This indicates that optimization of the source term can decrease the area of a repository remarkably, and the application of average source term is an efficient way in a deep geological disposal system design. In this paper, effectiveness of source term optimization was judged on the basis of spent fuels discharged from Kori unit 4. If entire spent fuels from all life-time of existing reactors are considered, the quantitative benefit will be changed. This conclusion, however, will not be changed, because decay heat considering burnup and cooling time of respective spent fuel will be definitely lower than that of reference spent fuel with conservativeness. 4. Conclusions Spent fuels to be disposed of have various initial 235U enrichments, discharge burnups, and cooling times, and consequently each spent fuel has a different decay heat. Because a disposal system design considering the respective spent fuel characteristics was previously not possible in Korea, a system was developed based on a conservative reference spent fuel nominated through historical data analyses on the relation of the spent fuel inventory with the initial 235U enrichment, discharge burnup, and cooling
time. Recently, ASOURCE, an advanced source-term calculation system, was developed by the Korea Atomic Energy Research Institute, making it possible to design a disposal system considering the respective source terms of spent fuels. In this study, the benefits resulting from consideration of the individual decay heat for entire spent fuels were evaluated in terms of the footprint of a repository. As a result, it was seen that conservativeness in the disposal system design can be remarkably avoided, revealing about a roughly 50% decrease in the footprint of repository, considering the respective irradiation and cooling profile. Therefore, it was concluded that optimization of the source term is a viable option in a deep geological disposal system design. Acknowledgements We would like to acknowledge that this work was funded by the Ministry of Science, ICT and Future Planning. References Cho, D.K., Choi, J.W., Hahn, P.S., 2006. Current status and projection of spent nuclear fuel for geological disposal system design. J. Kor. Radioactive Waste Soc. 4 (1), 87–93. Cho, D.K., Lee, Y., Lee, J.Y., Choi, J.W., 2007. Characteristics of a geological disposal system for the increasing burn-up of spent nuclear fuel in Korea. J. Nucl. Sci. Technol. 44 (10), 1306–1316. Cho, D.K., Kook, D.H., Choi, J.W., Park, J.H., 2013a. Advanced hybrid analysis system for nuclear facility design with best estimate source terms. Nucl. Eng. Des. 256 (2013), 274–284. Cho, D.K., Choi, H.J., Jeong, J., 2013b. Verification of source term estimation method against measured data for spent fuel hardware characterization. Ann. Nucl. Energy 58 (2013), 36–42. Croff, G., 1980. A User’s Manual for the ORIGEN2 Computer Code, ORNL/TM-7175, Oak Ridge National Laboratory Report. Gauld, I.C., Bowman, S.M., Horwedel, J.E., 2009. ORIGEN-ARP: Automatic Rapid Processing for Spent Fuel Depletion, Decay, And Source Term Analysis, ORNL/ TM-2005/39, Version 6, Vol. I, Section, D1. Oak Ridge National Laboratory. Korea Atomic Energy Research Institute, 2011. Geological Disposal of Pyroprocessed Waste from PWR Spent Nuclear Fuel in Korea. KAERI/TR-4525/2011. Lee, J.Y., Cho, D.K., Choi, H.J., Choi, J.W., 2007. Concept of a Korean reference disposal system for spent fuels. J. Nucl. Sci. Technol. 44 (12), 1565–1573.